Status of the JENDL Project
Osamu IWAMOTO and Kenji YOKOYAMA Japan Atomic Energy Agency
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Japanese Nuclear Data Committee Japan Atomic Energy Agency
chaired by K. Nakajima, Kyoto University Subcommittee on Nuclear Data (H.Harada, JAEA)
ENSDF Group (H.Iimura, JAEA) Japanese Nuclear Data Measurement Network (Y.Watanabe,
Kyushu Univ.) Activation Cross Section Evaluation WG (N.Iwamoto, JAEA)
Subcommittee on Reactor Constants (K. Okumura, JAEA)
Reactor Integral Test WG (G.Chiba, Hokkaido Univ.) Shielding Integral Test WG (C.Konno, JAEA) WG on Evaluation of Nuclide Generation and Decay Heat (F.
Minato, JAEA) Covariance Utilization WG (T.Iwasaki, Tohoku Univ.) Nuclear Data Processing Program WG (K.Suyama, JAEA)
Subcommittee on International Strategy (T. Fukahori, JAEA)
International Strategy WG (K.Suyama, JAEA)
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Plan of JENDL-5 • Improve reliability and completeness of data for various
applications • Revise data of light nuclei, structural material, FP, actinide • Include all stable isotopes • Increase covariance data • Add isomer production data for activation
• Use new R-matrix code, AMUR, for light nucleus
evaluation • Perform evaluation with new J-PARC data (mainly for MA) • Plan new simultaneous evaluation of fission cross section
for major actinide
• To be released in FY 2021
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Schedule of JENDL-5 development
Fiscal Year
2015 2016 2017 2018 2019 2020 2021
FP
Light nucleus
Structure mat.
Actinide
Covariance
S(α,β)
Libraries JENDL/PD-2016 JENDL/AD-2017
JENDL-5α JENDL-5β JENDL-5
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Light nucleus • New R-matrix code AMUR • Preliminary evaluation of O-16, F-19 • Covariance
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Analysis of 17O System
You can see traces of the (near) unitary limit
Re-normalization found to be ~1.0 (uncertainty < 0.5%)
16O(n,tot)
Correlation matrix Shape analysis
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Analysis of 19F neutron cross section Reich-Moore approximation
19F(n,tot) Expt. : Larson+ (1976)
Expt. : Guber+ (2005) 19F(n,γ)
Preliminary result
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Structural material • Add isomer production data • Update resonance parameters • Above the resonance region
• model calculation with CCONE • covariance data will be evaluated CCONE + KALMAN
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Evaluation of Nb
0 5 10 15 200
0.5
1
1.5
2
2.5
93Nb (n,n')
Incident Neutron Energy (MeV)
Cro
ss S
ectio
n (b
)
JENDL–4.0ENDF/B–VII.1Present
92 Simakov+94 Lashuk+
77 Kozur+74 Birjukov+73 Vanheerden+
63 Glazkov63 Thomson70 Goebel+
93 Wagner+ (m.s.)96 Ikeda+ (m.s.)
88 Wagner+ (m.s.)88 Gayther+ (m.s.)81 Ryves+ (m.s.)80 Taylor+ (m.s.)JENDL/A–96 (m.s.)ENDF/B–VII.1 (m.s.)JEFF–3.1/A (m.s.)Present (m.s.)
10 15 20 250
0.5
1
1.5
93Nb (n,2n)
Incident Neutron Energy (MeV)
Cro
ss S
ectio
n (b
)
JENDL–4.0ENDF/B–VII.1
80 Frehaut+84 Lychagin+
80 Prokopets77 Veeser+77 Kozur+
72 Mather+76 Holub+
99 Filatenkov+ (m.s.)00 Fessler+ (m.s.)
93 Ikeda+ (m.s.)91 Molla+ (m.s.)91 Smith+ (m.s.)90 Santry+ (m.s.)
JENDL/A–96 (m.s.)
00 Kiraly+ (m.s.)
11 Honusek+ (m.s.)07 Mannhart+ (m.s.)
90 Kimura+ (m.s.)
Present (m.s.)Present
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Evaluation of Cu isotopes
6/12
63Cu(n,n0) 63Cu(n,n1+2+3)
65Cu(n,n1+2+3)
Elastic and inelastic scattering angular distribution
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Various reaction cross sections on Cu
8/12
63Cu(n,2n)
63Cu(n,α)
63Cu(n,p)
65Cu(n,2n)
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Fission products and medium heavy nuclides • Objectives
• Updating old FP evaluations • Providing activation cross sections for decommissioning of
LWRs • Evaluated nuclides
• Ga, Nb, Tc, Sb, Te, I, Er • Ta, Pt, Hg, Tl • Most of them have already published from JNST.
• Evaluations with new experimental data at J-PARC are in progress.
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Evaluation of 99Tc 99Tc neutron capture cross section
Neutron energy (keV)
new data@J-PARC
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14 Resonance analysis of capture cross section of Gd measured by J-PARC ANNRI
Absolute capture cross section measurement by Kimura et al. Relative cross section for higher energy. A negative resonance for 157Gd.
157Gd
Actinides • Minor actinides such as Am-241, 243, Np-237 will be updated using new experimental data of J-PARC.
• Activation cross section data of MA were revisited. • Simultaneous evaluation for fission cross section of major actinides is planned.
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Current status of the thermal neutron capture cross section of MAs (e.g. 241Am)
Large discrepancy among the reported data of MAs.
The data measured
by activation were overestimated maximally by 20% from JENDL-4.0.
Westcott+Cd-ratio Averaged Cross section TOF
Activation
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R/𝜎𝜎0 = g𝜙𝜙1 + 𝑠𝑠0𝐶𝐶𝜙𝜙2 + 𝛿𝛿𝑠𝑠0𝜙𝜙2 R′/𝜎𝜎0 = g𝜙𝜙1′ + 𝑠𝑠0𝑐𝑐𝜙𝜙2′
Correction of Westcott convention for MAs
𝛿𝛿𝑠𝑠0 =2
𝜎𝜎0 𝜋𝜋�
𝑑𝑑𝑑𝑑𝑑𝑑 𝜎𝜎 𝑑𝑑 − g𝜎𝜎0
𝑑𝑑0𝑑𝑑
𝐸𝐸𝐶𝐶
𝜇𝜇𝑘𝑘𝑘𝑘
241Am Non-covered Reaction rate (R)
241Am Cd-covered Reaction rate (R’)
Transmission by Cd-filter
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Corrected thermal capture cross section for 241Am
Before correction After correction
Discrepancy among the data became smaller.
Westcott+Cd-ratio Averaged Cross section TOF
Activation M
ugha
ghab
(2
006)
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Summary of nuclear data evaluation • JENDL-5 is under development. • Evaluations for light nuclei, structure materials, fission products, and actinides are in progress.
• CIELO data will be tested with JENDL. • Thermal scattering law data of H2O will be updated in cooperation with Kyoto Univ.
• JENDL-5 will be released in FY 2021.
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Development of nuclear data processing system FRENDY • JAEA started developing a new nuclear data processing system FRENDY in 2013 • FRom Evaluated Nuclear Data librarY to any application • To process the nuclear data library by JAEA’s nuclear
application codes users • The first goal is processing the nuclear data for continuous energy Monte Carlo codes • For MVP, PHITS of JAEA and MCNP of LANL
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Nuclear data processing
system FRENDY
Nuclear data library
k-eff , flux, … Nuclear
application code MVP, MARBLE2,
PHITS K. Tada, Y. Nagaya, S. Kunieda, K. Suyama, and T. Fukahori, “FRENDY: A New Nuclear Data Processing System being Developed at JAEA,” ND2016
Features of FRENDY 21
• Utilization of modern programming techniques • C++, BoostTest library, Git • Improvement of quality and reliability
• Consideration of maintainability, modularity, portability and flexibility • Encapsulate all classes • Minimize the function • Maintain the independence of each module
• Processing methods of FRENDY is similar to NJOY99 • The modification of the processing methods and implementation of
the original functions will be investigated in the future • Reflecting requests of nuclear data processing system users
• Development of FRENDY is supported by many organizations and companies in Japan
Comparison for integral experiments • keff values of ICSBEP benchmark are compared
• MCNP sample input files in ICSBEP hand book are used • keff values of FRENDY are similar to that of NJOY99
• Differences are not so varied with the neutron spectra and the major fissile materials
• FRENDY properly generates ACE files
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0.99950
0.99975
1.00000
1.00025
1.00050 1σ : NJOY99
1σ : FRENDY
FREN
DY
/ NJO
Y99
Benchmarking toward Next JENDL • The 1st phase task of RIT-WG (Reactor Integral Test Working Group) in JENDL committee has been completed • Development of a comprehensive and ready-to-use standard
benchmark set, based on Japanese Monte-Carlo code MVP, for LWR nuclear data by utilizing open and well-evaluated integral experiments, such as ICSBEP and IRPhEP (NEA-DBs)
• New thick (152 pages) report was written up • RIT-WG, “Compilation of the Data Book on Light Water Reactor
Benchmark to Develop the Next Version of JENDL – Utilization of Criticality Data in ICSBEP and IRPhEP Open Databases,” JAEA-Data/Code 2017-006 (2017) [in Japanese]
• Two benchmark sets are recommended in the report 1. Light-water-moderated low-enriched U cores in NEA-DBs 2. Light-water-moderated MOX cores in NEA-DBs
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Recommended Benchmark Set for Light-water-moderated Low-enriched U cores
No. Objectives of Benchmark Series of Experiments Case Number Comments
LCT-001 (PNL) 1~8 U enrichment: 2.35wt%LCT-005 (PNL) 1, 5 , 14 U enrichment: 2.35, 4.31wt%LCT-006 (TCA) 1~18 U enrichment: 2.60wt%
LCT-048 (DIMPLE) 1~5 U enrichment: 3.01wt%LCT-002 (PNL) 1~5 U enrichment: 4.31wt%
LCT-079 (Sandia) 1 ,2 , 6 , 7 U enrichment: 4.31wt%LCT-007 (Vulduc) 1~10 U enrichment: 4.74wt%LCT-026 ( IPPE) 1, 3 U enrichment: 4.92wt%
LCT-018 (DIMPLE) 1 U enrichment: 7.00wt%
LCT-008 (B&W) 1~17 Boron concentration: 779~1511ppmKRITZ-2:1 1 Boron concentration: 218ppm
KRITZ-2:13 1 Boron concentration: 452ppm
Dependency on Gd concentration LCT-005 (PNL) 1~11, 14 Gd concentration: 0~1.48g Gd/L
Dependency on Rh inventory LCT-079 (Sandia) 1~10 Rh foil thickness(31 sheets): 0~100μm
LCT-026 ( IPPE) 1~4 Core temperature: 20, 231, 19, 206℃KRITZ-2:1 1~2 Core temperature: 20, 249℃
KRITZ-2:13 1~2 Core temperature: 22, 280℃
LCT-010 (PNL) 1~4, 20~23 U enrichment: 4.31wt%LCT-017 (PNL) 1~3, 23~25 U enrichment: 2.35wt%
LCT-010 (PNL) 5, 7~8, 24~30 U enrichment: 4.31wt%LCT-017 (PNL) 4~9, 26~29 U enrichment: 2.35wt%
LCT-010 (PNL) 9~19 U enrichment: 4.31wt%LCT-017 (PNL) 10~22 U enrichment: 2.35wt%
LCT-005 (PNL) 12, 13 , 16 Fuel pin pitch: 1.598cmLCT-005 (PNL) 15 Outliner of C/E by JENDL-4.0LCT-010 (PNL) 6 Outliner of C/E by JENDL-4.0
LCT-019 (Kurchatov) 1~3 Outliner of C/E by JENDL-4.0LCT-022 (Kurchatov) 1~7 Outliner of C/E by JENDL-4.0LCT-024 (Kurchatov) 1~2 Outliner of C/E by JENDL-4.0LCT-025 (Kurchatov) 1~4 Outliner of C/E by JENDL-4.0
Reflector effect (Fe)
Total: 24 data8
Unrecommendedfor benchmarking
Total: 21 data
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Reflector effect (Depleted U)
Total: 20 data7
Total: 8 data
Dependency on core temperature5
Total: 14 data
Ref lector effect (Pb)6
Total: 12 data3
Total: 10 data4
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U-235, U-238 Basic Benchmarks
Dependencies on - U enrichment - H/U atomic number ratio - Fuel pin pitch - Number of fue l pins
Total: 56 data
Total: 19 data
2
Dependency on Boronconcentration
(Room temperature)
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Result of Recommended Benchmark Set for Light-water-moderated Low-enriched U cores
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0.990
0.995
1.000
1.005
1.010
0 2 4 6 8
LCT-001(PNL)LCT-005(PNL)LCT-006(TCA)LCT-018(DIMPLE)LCT-026(IPPE)LCT-048(DIMPLE)LCT-002 (PNL) LCT-079(Sandia)
C/E
val
ue
U enrichment
±0.4%
Fig. Dependency on U enrichment for Criticality (Benchmark Set No.1 with 56 data)
• Experiments using neutron absorbers and FPs (i.e., B, Gd, and Rh) are excluded • Only cores with H2O (not Pb, Depleted U, or Fe) reflectors in room temperature areselected • 18 data judged as outliners by the trend of C/E values are excluded
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Recommended Benchmark Set for Light-water-moderated MOX cores
No. Objectives of Benchmark Series of Experiments Case Number Comments
KRITZ-2:19 (Studsvik) 1 Pu enrichment: 1.5wt%, Pu240: 7.8at%MCT-009 (CAF) 2~6 Pu enrichment: 1.5wt%, Pu240: 7.8at%
MCT-002 (PRCF) 1, 3 , 5 Pu enrichment: 2.0wt%, Pu240: 7.7at%MCT-006 (CAF) 1~6 Pu enrichment: 2.0wt%, Pu240: 7.7at%MCT-007 (CAF) 1~5 Pu enrichment: 2.0wt%, Pu240: 16.5at%MCT-008 (CAF) 1~6 Pu enrichment: 2.0wt%, Pu240: 23.4at%
MCT-004 (TCA) 1~11Pu enrichment: 3.0wt%(low density),Pu240: 22.0at%
MCT-005 (CAF) 1, 2 , 4~7 Pu enrichment: 4.0wt%, Pu240: 18.1at%MCT-003 (CRX) 1, 2 , 4~6 Pu enrichment: 6.6wt%, Pu240: 8.5at%MCT-001 (CML) 1~4 Pu enrichment: 22.3wt%, Pu240: 11.4at%MCT011 (Valduc) 1~6 Pu enrichment: 25.6wt%, Pu240: 9.7at%
MCT-002 (PRCF) 1~6 Boron concentration: 688~1090ppmMCT-003 (CRX) 2, 3 Boron concentration: 337ppm
Dependency on core temperateure KRITZ-2:19 (Studsvik) 1~2 Core temperature: 21, 236℃
MCT-009 (CAF) 1 Outliner of C/E by JENDL-4.0MCT-005 (CAF) 3 Outliner of C/E by JENDL-4.0
3Total: 2 data
4Unrecommended for benchmarking
Total: 2 data
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Pu, U-238 basic benchmarks
Dependencies on - Pu enrichment - H/Pu atomic number ratio - Fuel pin pitch - Number of fue l pins
Total: 58 data
2Dependency on Boron
concentrationTotal: 8 data
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Result of Recommended Benchmark Set for Light-water-moderated MOX cores
0.990
0.992
0.994
0.996
0.998
1.000
1.002
1.004
1.006
1.008
1.010
0 200 400 600 800 1000 1200
KRITZ-2:19(Pu enrich. 1.5wt%、
7.8%Pu240) MCT009 -Battelle(Pu enrich. 1.5wt%、
7.8%Pu240) MCT002 -PNL(Pu enrich.2.0wt%、
7.7%Pu240) MCT006 -CAF(Pu enrich. 2.0wt%、
7.7%Pu240) MCT007 -CAF(Pu enrich. 2.0wt%、
16.5%Pu240) MCT008 -CAF(Pu enrich. 2.0wt%、
23.4%Pu240) MCT004 -TCA(Pu enrich. 3.0wt%、
22.0%Pu240) MCT005 -Hanford(Pu enrich. 4.0wt%、
18.1%Pu240) MCT003 -CRX(Pu enrich. 6.6wt%、
8.5%Pu240) MCT001 -PNL(Pu enrich. 22.3wt%、
11.4%Pu240) MCT011 -Valduc(Pu enrich. 25.6wt%、
9.7%Pu240) H/Pu atomic number
C/E
val
ue
±0.4%
Fig. Dependency on H/Pu atomic number ratio for Criticality (Benchmark Set No.1 with 58 data)
• Experiments using a large amount of Boron are excluded • Cores with high temperature H2O moderator are excluded • 2 data judged as outliners by the trend of C/E values are excluded
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Example of Sensitivity Analysis
(c) Contribution of Pu-239 fission to criticality of MCT-009_case6
F32-J40: -0.177%dk/kk’
B71-J40: -0.114%dk/kk’
• Sensitivity analysis has been started by using the recommended benchmark set
• Big difference of nuclear data does not always affect the integral parameters
• It is very useful to determine the important energy regions and reactions
(a) Differences of Pu-239 fission cross section compared with ENDF/B-VII.1 and
JEFF-3.2
0.004~0.4eV
(b) Sensitivity coefficient of Pu-239 fission for MCT-009_case6
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Development of automatic nuclear data validation system VACANCE • JAEA started developing an automatic nuclear data validation system VACANCE in 2016 • Validation Environment for Comprehensive and Automatic
Neutronics Calculation Execution
• VACANCE significantly reduces workload and human error • Automatically runs nuclear data
validation cycle
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K. Tada and K. Suyama, “Development of an Automatic Nuclear Data Validation System VACANCE”, ICAPP2017
Characteristics of VACANCE (1/2) • Automatic input file search
• VACANCE automatically searches all input files in the user specified directory
• Automatic input file modification • To easily change the input
option and nuclear data library • Number of history, batch size, flux
output option, … • JENDL → ENDF, JEFF, TENDL, …
• Enables a parallel computation using OpenMP • To efficiently run the hundreds of
calculations • Enable restart calculation
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Top
LCT01
HCT
LCT26
LCT07
VACANCE
VACANCE input
LCT01-1 .inp
LCT01-3 .inp …
LCT26-2 .inp
LCT26-1 .inp
…
LCT …
【 Example of directory structure of integral experiment calculations using VACANCE 】
Characteristics of VACANCE (2/2) • Drives two neutronics calculation codes
• MVP and MCNP • Extensibility is considered to easily add other calculation codes • FRENDY and MARBLE2 will be added in the near future
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• Makes comparison table of C/E value • If experimental value is
given, C/E value is automatically calculated
• Sort results by user-specified physical value • 235U/Pu Enrichment, H/U
ratio, temperature, and so on -0.4%
-0.2%
0.0%
0.2%
0.4%
2 4 6 8
C/E
-1
235U enrichment [wt%]
-0.8%
-0.4%
0.0%
0.4%J40 E8βF32
• Estimating the impact of the nuclear data difference • Comparison among calculation results using several
nuclear data library is important to validate nuclear data
Example 32
【 Comparison of C/E-1 value of keff changing nuclear data of 238U 】
C/E
-1
• VACANCE improves efficiency of this comparison • Many steps are required
to compare the calculation results • Modification of input files,
evaluation of k-eff, calculation of C/E value and edit results
• VACANCE automatically runs above steps with simple input file
Summary of nuclear data validation Validation process towards JENDL-5 is in progress
1. Development of nuclear data processing system, FRENDY, has been successfully started
2. Two recommended benchmark sets for LWR selected carefully from ICSBEP and IRPhEP have been determined
3. New automatic nuclear data validation system, VACANCE, will significantly reduce workloads and human errors in the validation process
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